1. NAME AND TITLE OF DATA LIBRARY
FSXLIB-J3: MCNP continuous energy neutron cross section library based on JENDL-3.
See DLC-190/FSX96 based on JENDL3.2.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
MAKXSF - Fortran 77 code for converting cross sections to binary form.
Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-11 Japan.
4. HISTORICAL BACKGROUND AND INFORMATION
In 1988, the first version of FSXLIB was released at Japan. It included 17 nuclides based on JENDL-3PR2, ENDF/B-IV and ENDF/B-V.
In 1990, the second version of FSXLIB was informally released at Japan. It included 68 nuclides and added 48 nuclides based on JENDL-3 (revision-0) to the first version.
In 1992, the FSXLIB-J3 was released. It was based on JENDL-3 (revision-1).
5. APPLICATION OF THE DATA
FSXLIB-J3 may be used as a continuous energy cross section table of neutron interaction for the MCNP code.
6. SOURCE AND SCOPE OF DATA
The original evaluated nuclear data file is JENDL-3 (revision-1) released in the end of 1990. The processing code which produced the pointwise cross sections is the modified version of nuclear data processing system NJOY (version 83/6).
The number of nucides included in FSXLIB-J3 is 116 from JENDL-3 (revision-1).
The format of FSXLIB-J3 conforms with the continuous energy neutron cross section table as defined in the MCNP manual. The photon production data is processed into an "expanded format." The processing temperature of cross section is 300 K.
The energy ranges for neutron and photon are 10-5 eV to 20 MeV and 10 keV to 20 MeV depends on JENDL-3, respectively.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAMS
MAKXSF is a program which will convert MCNP Type 1 data to Type 2 (binary).
8. DATA FORMAT AND COMPUTER
BCD card images; FACOM M-780; IBM RISC 6000; any UNIX-based computer.
9. TYPICAL RUNNING TIME
The typical running time for MACROS is less than a minute on FACOM M-780.
a. included in the documentation:
K. Kosako, Y. Oyama, and H. Maekawa, "FSXLIB-J3: MCNP Continuous Cross Section Library Based on JENDL-3," Proceedings of New Horizons in Radiation Protection and Shielding, (1992).
b. background information:
K. Kosako, Y. Oyama, and H. Maekawa, "FSXLIB-J3: MCNP Continuous Cross Section Library based on JENDL-3," JAERI-M 91-187 (1991) (in Japanese).
K. Shibata, et. al., "JENDL-3: Japanese Evaluated Nuclear Data Library, Version-3," JAERI- 1319 (1990).
J. F. Briesmeister (editor), "MCNP - A General Monte Carlo Code for Neutron and Photon Transport," LA-7396-M, revised 2 (1986).
R. E. MacFarlane, D. W. Muir, and R. M. Boicourt, "The NJOY Nuclear Data Processing System," LA-9303-M (ENDF-324) (1982).
11. CONTENTS OF LIBRARY
Included are the referenced document, retrieval code and libraries.
12. DATE OF ABSTRACT
KEYWORDS: BASED ON JENDL; EVALUATED NEUTRON CROSS SECTIONS; MCNP FORMAT; NEUTRON CROSS SECTIONS